Effect of helium on void formation in nickel

This study examines the influence of helium on void formation in self-ion irradiated nickel. Helium was injected either simultaneously with, or prior to, the self-ion bombardment. The void microstructure was characterized as a function of helium deposition rate and the total heavy-ion dose. In particular, at 575°C and 5 × 10^?3 displacements per atom per second the void density is found to be proportional to the helium deposition rate. The dose dependence of swelling is initially dominated by helium driven nucleation. The void density rapidly saturates after which swelling continues with increasing dose only from void growth. We conclude that helium promotes void nucleation in nickel with either helium implantation technique, pre-injection or simultaneous injection. Qualitative differences, however, are recognized.     

The temperature dependence of nickel-ion damage in nickel

The temperature dependence of void and dislocation structures was studied in high-purity nickel irradiated with 2.8 MeV ⁵⁸Ni⁺ ions to a displacement density of 13 displacements per atom (dpa) at a displacement rate of $\text{7 × 10}{}^{\mathrm{?2}}\text{dpa/sec}$ over the temperature range 325 to 625°C. Dislocation loops, with no significant concentrations of voids, were observed in specimens irradiated at 475°C and below. Specimens irradiated between 525 and 725°C contained both voids and dislocations. The maximum swelling was measured as 1.2% at 625°C. Analysis of the data by theoretical models for void nucleation and growth indicated that the swelling in the present experiment was principally limited by void growth at low temperatures and by void nucleation at high temperatures. The data were also compared with previously reported neutron and nickel-ion irradiation results.     

Numerical simulation of void nucleation in irradiated metals

An atomistics-based theory for void nucleation has been used to calculate — for the first time — terminal void number densities for irradiated nickel, type 316 stainless steel, and the Ni-base alloy, PE-16. Both the absolute magnitudes and temperature dependences of the void number densities are in agreement with experiment. The void nucleation parameter, ψ, which governs spontaneous void nucleation was evaluated for the three materials; the results are in agreement with experiment. The critical gas content for rapid void growth was calculated for PE-16 and type 316 stainless steel, and was found to increase from about 10 helium atoms at the lower end of the void swelling range to some 10⁴ atoms at the upper end. The theory was also found to predict re-nucleation of a new distribution of voids after a drop in temperature during irradiation.     

Cavity nucleation calculations for irradiated metals

A theory of helium-assisted cavity nucleation in irradiated metals is modified and applied to conditions of continuous helium generation. The theory considers the nucleation and growth of cavities by coprecipitation of vacancies, interstitials, and inert gas atoms. Calculations are performed for type 304 stainless steel for comparison with ion irradiation experiments at $\text{~ 2 × 10}{}^{\mathrm{?4}}\text{dpa/s}$, with helium implantation at the rate of ~10^?2 appm/s, to a total damage of ~ 5 dpa, over the temperatures 773–973 K. Total cavity number density calculated ranges from 10²³ m^?3 at 773 K to 10²⁰ m^?3 at 973 K. The calculated incubation time for cavity appearance is 1000–3000 s (0.2–0.6 dpa). The calculated plot of cavity density versus time approximately reproduces the experimental data. Predicted cavity size distributions are roughly bell-shaped, but skewed in favor of larger cavity sizes. Calculated and experimental mean sizes agree within a factor of 3. The predictions of the model are found to change very little when most parameters are varied within reasonable limits. The model is, however, found to be strongly sensitive to cavity : matrix surface energy, as well as the rate that helium atoms are displaced from dislocations.     

The effect of damage rate on void growth in metals

The rate theory formulation of void growth was utilized to analyze the effects of damage rate on metal swelling. In particular, the swelling behavior of 316 SS was modeled as a function of temperature, over a range of displacement damage rates between 10^?6dpa/s and 10^?3dpa/s. Detailed analysis of the rate processes for point defect annihilation, migration, and loss to sinks indicated that small vacancy loops limit void growth at high damage rates. The reduction of void growth rates by vacancy emission from voids was found to be shifted towards higher temperature at higher displacement rates. In effect, the peak swelling temperature as well as the upper cutoff temperature for swelling are increased as the displacement rate is increased. The influence of constant or rate dependent nucleation conditions on the final swelling was investigated and it was shown that the initial microstructure before the growth stage essentially determines the peak swelling temperature. When appropriate empirical expressions for void and loop densities were used, the final peak swelling temperature shift agrees reasonably well with experimental data.     

The influence of He/dpa ratio and displacement rate on microstructural evolution: a comparison of theory and experiment

A kinetic model was developed to investigate the influence of the displacement rate and helium generation rate on microstructural evolution in austenitic stainless steels. The model integrates the rate equations describing the evolution of point defects, small point defect clusters, helium-vacancy clusters, and the larger cavity size distribution that is responsible for observable swelling. Cavity (bubble) nucleation is accounted for by the helium-vacancy cluster evolution, while void formation occurs when bubbles grow beyond a critical size in the larger cavity distribution.

A series of ion irradiation experiments were used to both calibrate the model and to provide a comparison between model predictions and experimental observations. The experiments involved single and dual-beam irradiations of solution annealed AISI-316 stainless steel at 873 K. The displacement rates were in the range of 2 × 10⁻³ to 1 × 10⁻² dpa/s and the helium-to-dpa ratios were in the range of 0 to 50 appm He/dpa. The maximum displacement dose was 25 dpa. The experiments revealed a significant effect of helium on both the dislocation structure and the cavity distribution. The model predictions of helium effects over a broad range of He/dpa ratios and displacement rates were consistent with experimental observations.     

Observation of a partially-ordered void lattice in aluminium irradiated with 400 keV Al+ ions

Transmission electron microscopy observations of voids formed in aluminium during irradiation at 50°C and 75°C with 400 keV Al⁺ ions, have shown that partially-ordered void arrays are often present. These arrays occur in high-purity annealed aluminium, which has been implanted with 10^?4 atom/atom helium before ion irradiation. The void concentration is found to be $\text{～3 × 10}{}^{16}\text{/ cm}{}^3$, and the void lattice parameter ～ 700 Å. The ratio of void lattice parameter to void radius is ～ 12. Ordered void lattices have been observed frequently in irradiated body-centred cubic metals but the only previous observation for a face-centred cubic metal was in nickel. Theoretical predictions of void lattices in metals are discussed and related to the observations reported herein.     

Interstitial loop nucleation and growth in solution-treated type 316 stainless steel irradiated to low doses with 22 MeV C2+ and 46.5 MeV Ni6+ ions

The nucleation and growth of interstitial dislocation loops have been studied in solution-treated type 316 austenitic steel irradiated to low doses in the Harwell Variable Energy Cyclotron. Specimens have been irradiated with 46.5 MeV Ni⁶⁺ ions and 22 MeV C²⁺ ions, after room temperature pre-injection with 10 ppm helium and without helium pre-injection, at temperatures in the range 300–600°C. The effects of these irradiation variables on the interstitial loop populations produced are discussed. At low doses, where loop intersection is rare and dislocation network formation is minimal, the number of interstitial atoms stored in loops can give an indication of the swelling rate in circumstances where voids remain submicroscopic. It is shown that extrapolation of the low-dose swelling rates indicated by interstitial loop populations gives reasonable fit with experimentally determined high-dose void swelling values.     

55Fe effect on enhancing ferritic steel He/dpa ratio in fission reactor irradiations to simulate fusion conditions

This study evaluated methods for increasing the helium production rate in ferritic steel irradiation in a fission reactor neutron spectrum in order to increase the helium to atomic displacement ratio to values typical of fusion reactor first wall conditions. An early experiment showed that the accelerated He(appm)/dpa ratio of about 2.3 was achieved for 96% enriched ⁵⁴Fe in iron in the High Flux Isotope Reactor (HFIR), ORNL. In the current work, the ferritic steel He(appm)/dpa ratio was studied in the neutron spectrum of HFIR with the ⁵⁵Fe thermal neutron helium production taken into account. A benchmark calculation for the same sample, as used in the aforementioned experiment, was then used to adjust and evaluate the ⁵⁵Fe (n, a) cross section values in TALYS-based Evaluated Nuclear Data Library (TENDL). The analysis showed that a decrease of a factor of 6700 for the TENDL ⁵⁵Fe (n, a) cross section in the intermediate and low energy regions was required in order to fit the experimental results. The best fit to the cross section value at thermal neutron energy was about 27 mb. With the adjusted ⁵⁵Fe (n, a) cross sections, calculation showed that the ⁵⁴Fe and ⁵⁵Fe isotopes could be enriched by the isotopic tailoring technique in a ferritic steel sample irradiated in HFIR to significantly enhance the helium production rate. This new calculation can be used to guide future isotopic tailoring experiments designed to increase the He(appm)/dpa ratio in fission reactors. A benchmark experiment is suggested to be performed to evaluate the ⁵⁵Fe (n, a) cross section at thermal energy.     

Proton irradiation emulation of PWR neutron damage microstructures in solution annealed 304 and cold-worked 316 stainless steels

Solution annealed (SA) 304 and cold-worked (CW) 316 austenitic stainless steels were pre-implanted with helium and were irradiated with protons in order to study the potential effects of helium, irradiation dose, and irradiation temperature on microstructural evolution, especially void swelling, with relevance to the behavior of austenitic core internals in pressurized water reactors (PWRs). These steels were irradiated with 1 MeV protons to doses between 1 and 10 dpa at 300 °C both with or without 15 appm helium pre-implanted at ∼100 °C. They were also irradiated at 340 °C, but only after 15 appm helium pre-implantation. Small heterogeneously distributed voids were observed in both alloys irradiated at 300 °C, but only after helium pre-implantation. The pre-implanted steels irradiated at 340 °C exhibited homogenous void formation, suggesting effects of both helium and irradiation temperature on void nucleation. Voids developed sooner in the SA304 alloy than CW316 alloy at 300 and 340 °C, consistent with the behavior observed at higher temperatures (>370 °C) for similar steels irradiated in the EBR-II fast reactor. The development of the Frank loop microstructure was similar in both alloys, and was only marginally affected by pre-implanted helium. Loop densities were insensitive to dose and irradiation temperature, and were decreased by helium; loop sizes increased with dose up to about 5.5 dpa and were not affected by the pre-implanted helium. Comparison with microstructures produced by neutron irradiation suggests that this method of helium pre-implantation and proton irradiation emulates neutron irradiation under PWR conditions.     

Effect of Fast Neutron Irradiation on the Properties of Boron Carbide Pellet

Boron carbide pellets were irradiated in the experimental fast reactor “JOYO” to ¹⁰B burnup of up to 170x10²⁶cap/m³, fluences of 2x10²⁶/m²(E>0.1MeV), and maximum temperatures of about 1,200°C. Post irradiation examinations were made of microstructural changes, helium release, swelling, and thermal conductivity.

Boron carbide pellets irradiated to high burnups developed extensive cracking. Helium release from the pellets was initially low, but enhanced helium release was observed at high burnups and high temperatures. The swelling linearly increased with burnup, and when boron carbide was irradiated at high temperatures, the swelling rate began to decrease corresponding to the beginning of enhanced helium release. The correlation between swelling and the helium release was studied and the swelling was interpreted in terms of accumulation of helium in the boron carbide pellet. The thermal conductivity of the boron carbide pellets decreased rapidly by neutron irradiation accompanied with loss of temperature dependence.     

Development of fatigue life criteria for experimental fusion reactor first-wall structures

An approach to the rational design of fusion reactor first-wall structures against fatigue crack growth is proposed. The approach is motivated by microstructural observations of fatigue crack growth enhancement in unirradiated materials due to volumetric damage ahead of a propagating crack. Examples are cited that illustrate the effect of mean stress on void nucleation and coalescence, which represent the dominant form of volumetric damage at low temperature, and of grain boundary sliding and creep cavitation, which are the dominant volumetric damage mechanisms at high temperature. The analogy is then drawn between these forms of fatigue crack growth enhancement and those promoted by irradiation exposure in the fusion reactor environment, such as helium embrittlement and atomic displacement. An enhanced strain range is suggested as a macroscopic measure of the reduction in fatigue life due to the higher fatigue crack growth rates. The enhanced strain range permits a separation of volumetric and cyclic effects, and assists in the assignment of rational design factors to each effect. A series of experiments are outlined which should provide the numerical values of the parameters for the enhanced strain range.     

Effect of helium on void swelling in molybdenum

Simultaneous irradiation of molybdenum with helium and heavy ions (Ta³⁺) using a dual beam facility resulted in continued void nucleation in molybdenum to high dose levels, but the added helium had no measurable effect on the void swelling or swelling rate when compared with results for heavy ion irradiation without helium. Pretreatment by neutron irradiation or preinjection with helium resulted in no significant microstructural changes compared to no pretreatment. Also the temperature dependence of swelling was essentially unchanged when helium was added to the irradiation. The lack of a strong helium effect was attributed to the high inherent void nucleation rate in molybdenum. The overall swelling rate was similar to that observed for neutron irradiation and correlated well with the microstructural features that were observed. At the highest temperature and dose (1475 K and 40 dpa), simultaneous helium and heavy ion irradiation did result in a very nonuniform void distribution; thus, helium may have a greater effect on the microstructure at temperatures above those reported here.     

Positron annihilation study and computational modeling of defect production in neutron-irradiated reactor pressure vessel steels

Positron annihilation spectroscopy (PAS) and a computer simulation were used to investigate a defect production in reactor pressure vessel (RPV) steels irradiated by neutrons. The RPV steels were irradiated at 250 °C in a high-flux advanced neutron application reactor. The PAS results showed that mainly single vacancies were created to a great extent as a result of a neutron irradiation. Formation of vacancies in the irradiated materials was also confirmed by a coincidence Doppler broadening measurement. For estimating the concentration of the point defects in the RPV steels, we applied computer simulation methods, including molecular dynamics (MD) simulation and point defect kinetics model calculation. MD simulations of displacement cascades in pure Fe were performed with a 4.7 keV primary knock-on atom to obtain the parameters related to displacement cascades. Then, we employed the point defect kinetics model to calculate the concentration of the point defects. By combining the positron trapping rate from the PAS measurement and the calculated vacancy concentrations, the trapping coefficient for the vacancies in the RPV steels was determined, which was about 0.97 × 10¹⁵ s⁻¹. The application of two techniques, PAS and computer simulation, provided complementary information on radiation-induced defect production.     

Weldability of neutron irradiated austenitic stainless steels

Degradation of weldability in neutron irradiated austenitic stainless steel is an important issue to be addressed in the planning of proactive maintenance of light water reactor core internals. In this work, samples selected from reactor internal components which had been irradiated to fluence from 8.5 × 10²² to 1.4 × 10²⁶ n/m² (E > 1 MeV) corresponding to helium content from 0.11 to 103 appm, respectively, were subjected to tungsten inert gas arc (TIG) welding with heat input ranged 0.6–16 kJ/cm. The weld defects were characterized by penetrant test and cross-sectional metallography. The integrity of the weld was better when there were less helium and at lower heat input. Tensile properties of weld joint containing 0.6 appm of helium fulfilled the requirement for unirradiated base metal. Repeated thermal cycles were found to be very hazardous. The results showed the combination of material helium content and weld heat input where materials can be welded with little concern to invite cracking. Also, the importance of using properly selected welding procedures to minimize thermal cycling was recognized.     

Defect structure of neutron irradiated boron carbide

Boron carbide irradiated in thermal and fast reactor neutron spectra at 500 to 840°C up to 81% ¹⁰B burnup has been examined by transmission electron microscopy. Helium bubbles formed by agglomeration of helium from the ¹⁰B (n,α)⁷ Li reaction constitute the only visible form of irradiation-induced damage. The helium bubbles tend to have a disc-like morphology with the plane of the disc parallel to (111). Strong lattice strain fields around the bubbles evidenced a lack of thermal equilibrium. Bubble nucleation was influenced markedly by grown-in dislocations and stacking faults. Bubble coalescence, either by impingement or enhanced by bubble stress fields frequently gave rise to both trans- and intergranular microcracking. The significance of these observations on gas release and swelling of boron carbide are discussed briefly.     

Irradiation of U3Si-based compounds in the high-voltage electron microscope

U₃Si and U-3.5 wt% Si-1.5 wt% Al have been irradiated in a high-voltage electron microscope (HVEM) at 500–1180 keV and 300–660 K. The creation of ‘black spot’ damage and removal of deformation twins are observed. Defects about 10 nm diameter averaging 4 × 10²¹ m^?3 are seen only in samples pre-injected with 10^?5 atomic fraction argon and are tentatively identified as voids. The atomic displacement rate during HVEM irradiation of U₃Si-based compounds is about two orders of magnitude higher than that for fuel at power reactor ratings. It is inferred that displacement of silicon, and possible uranium, atoms in U₃Si-based compounds occurred in the HVEM at accelerating voltages in the range 700–1180keV.     

Analysis of Tritium Behavior in Very High Temperature Gas-Cooled Reactor Coupled with Thermochemical Iodine-Sulfur Process for Hydrogen Production

The tritium concentration in the hydrogen product in Japan's future very high temperature gas-cooled reactor (VHTR) system coupled with a thermochemical water-splitting iodine-sulfur (IS) process (VHTRIS system), named GTHTR300C, was estimated by numerical analysis. The tritium concentration in the hydrogen product significantly depended on undetermined parameters, i.e., the permeabilities of a SO₃ decomposer and a H₂SO₄vaporizer made of SiC. Thus, the estimated tritium concentration in the hydrogen product for the conservative analytical condition ranged from 3.4 × 10^?3 Bq/cm³ at STP (38 Bq/g-H₂) to 0.18 Bq/cm³ at STP (2,000 Bq/g-H₂). By considering the tritium retained by core graphite and the reduction in permeation rate by an oxide film on the heat transfer tube of the IHX and the HI decomposer, the tritium concentration in the hydrogen product decreased to the range from 3.3 × 10^?5 Bq/cm³ at STP (0.36 Bq/g-H₂) to 5.6 × 10^?3 Bq/cm³ at STP (63 Bq/g-H₂), which were smaller than those for the conservative analytical condition by factors of about 3.2 × 10^?2 and 9.6 × 10^?3, respectively. The effectof the helium flow rate in the helium purification system on the tritium concentration in the hydrogen product was also evaluated.     

Effects of neutron-induced ion implantation into blanket structural materials

The high flux of energetic neutrons in CTR blankets will lead to an appreciable implantation of light particles in structural materials as a consequence of neutron impacts. The influence of implanted helium and lithium on the material properties of CTR structural materials should be known because both media are prospective coolants. 30keV lithium ions were implanted into pre-thinned Nb foils. Electron micrographs showed black dots which were identified as precipitates of Li. A concentration profile of He in Nb, which suitably simulates the implantation profile in CTR's, was obtained with the help of the ¹⁰B(n, α)⁷Li reaction. Nb samples which were irradiated up to a He dose of 6 × 10¹⁶ cm^?2 showed heavily damaged surfaces after annealing. Surface erosion of CTR blanket materials due to this process cannot be excluded.     

Crack formation,crack healing and porosity redistribution during irradiation of UO2 and (U,Pu) O2

The irradiation-induced void volume redistribution in the fuel was analysed. The radial crack volume and porosity distributions, the central radii and the radial gap width were measured after irradiation and compared with the calculated values. Short-time (He-loop experiments in the FR2 reactor), medium-time (bundle irradiation in the BR2 reactor) and long-time (trefoil-irradiation in the DFR reactor) irradiated fuel pins were examined. The model of pore migration, used in the computer code SATURN-la, is based on the evaporation-condensation mechanism. Measured swelling rates were extrapolated to higher temperatures and used. The crack volume distribution was calculated on the basis of a multifractured fuel model. One can conclude from the comparison between calculated and measured void volume distributions that several mechanisms redistribute void volume. These are crack formation, crack healing, migration of sinter pores and fission gas bubbles, gas swelling, evaporation-condensation phenomena in the region of the central void, irradiation-induced sintering and increase in diameter of the cladding.